1. Field of the Invention
This invention relates to nuclear steam turbine power plant controls, and in particular, to a power plant control system which permits operation beyond the normally defined end of the fuel cycle of the reactor element.
2. Description of the Prior Art
In a nuclear reactor power plant, heat energy generated within the nuclear reactor element is utilized to provide a source of pressurized dry and saturated or low superheat steam. The steam so produced drives an associated turbine system linked mechanically to an electrical generator which produces electrical energy for an associated electrical load. In general, there are two main classifications of nuclear power plants, the Boiling Water Reactor or BWR System, and the Pressurized Water Reactor, or PWR System.
The Boiling Water Reactor, or BWR, conducts a water coolant through a single closed loop comprising the reactor element and the turbine system. As the water coolant passes through the reactor core, heat taken from the reactor core raises the temperature of the water coolant above the vaporization point of the coolant, thereby producing steam. The steam is then directly passed through the turbine system and the heat energy within the steam is converted into useful mechanical energy. The steam passes into a condenser element and is returned to a liquid state. The condensate, now in the form of low temperature water coolant, is reintroduced as a coolant into the reactor element, thus completing the single closed loop.
However, since the water coolant and the steam produced therefrom are exposed to the highly radioactive reactor core and have undergone extensive bombardment of high energy neutron particles therein, the turbine system through which the radioactively contaminated steam has passed experiences an increase in radioactivity levels. Also, extreme care must be taken to prevent leakage of the contaminated steam or water out of the single, closed BWR loop.
On the other hand, power plants employing a Pressurized Water Reactor use a double closed loop arrangement for the production of electrical energy by use of the nuclear reactor element. The first, or primary, side connects the reactor element in a closed loop with a steam generator, or heat exchanger element. The water coolant is maintained under a high pressure, approximately 2200 p.s.i. and takes heat from the reactor core as the pressurized water coolant passes therethrough. Since the water coolant is under pressure, the heat taken from the reactor does not raise the temperature of the water coolant above the vaporization temperature for water at the appropriate pressure.
Instead, the heated, but not boiling, water enters the steam generator element. In the steam generator, the heat carried by the pressurized water coolant is transferred to secondary water disposed within the steam generator. The transfer of heat from the pressurized water to the secondary water sufficiently cools the pressurized water coolant to permit its reintroduction into the reactor core, thus completing the primary side loop.
Within the steam generator, heat taken from the pressurized water coolant and transmitted to the secondary water therein raises the temperature of the secondary water above its boiling point, thus producing steam. However, in the case of the PWR, the steam produced within the steam generator element is not radioactively contaminated since the pressurized water coolant is physically segregated from the secondary water in the steam generator element.
From the steam generator, the steam produced by the heating of the secondary water is conveyed through the second enclosed loop, called the secondary or steam side. The secondary or steam side comprises a high pressure turbine element, a low pressure turbine element, a condenser element, and a series of feedwater heater elements. Each turbine element has an inlet orifice of the predetermined size with the inlet orifice of the high pressure turbine being physically smaller than the inlet orifice of the low pressure turbine element. Steam passes through the turbine elements and converts the heat energy carried therein to mechanical energy. The expanded steam is returned to the liquid state in the condenser. The condensate is then reintroduced into the steam generator after passing through a series of feedwater heater elements to complete the second closed loop.
It is evident then, that the pressure of the steam flowing in the secondary loop is directly dependent upon the amount of heat transferred from the pressurized water coolant to the secondary water in the steam generator. If the temperature of the water coolant is high, a greater amount of heat energy will be transferred to the secondary water than would occur if the primary water coolant were at a lower temperature. The heat energy transferred to the secondary water thus raises the temperature of a greater volume of secondary water above the applicable vaporization temperature of the secondary water, thus producing more steam in the secondary loop. Steam flow through the secondary or steam side loop is dependent upon the physical sizing of the high pressure turbine inlet orifice.
The heat content of the primary water coolant is obviously dependent upon the heat output in the reactor core. Heat output in the reactor core is in turn dependent upon the reactivity levels within the fuel elements which comprise the core.
Fuel element reactivity is dependent upon the age of the individual fuel elements. Early in the fuel element life, reactivity levels are usually above some predetermined reference point and controlled by the manipulation of a plurality of neutron absorbing control rods. If, for example, the reactivity level within the reactor element is desired to be raised, the control rods are withdrawn a predetermined distance from within the core, thus increasing the neutron level within the core, thereby increasing the reactivity level within the core. With the reactivity level within the core increased, heat is generated within the reactor and therefore the heat content of the pressurized water coolant increases.
A higher coolant temperature increases the amount of heat transferred from the pressurized water coolant to the secondary water in the steam generator. The increased amount of heat transferred increases the pressure of the steam produced in the steam generator for use in the secondary or steam side. The ability to change steam pressure to change the overall electrical output of the power plant to meet the demand of the system is termed the load follow capability of the system. As is well known to those skilled in the art, the load follow capability of the system is inhibited during a xenon transient condition.
The useful life, or fuel cycle, of the fuel elements utilized by the reactor is measured in terms of certain predetermined parameters. One such parameter is the amount of reactivity in excess of a certain predetermined reference level. This level of excess reactivity is above the reference level at the start of the fuel cycle, but the level of excess reactivity decreases toward the reference value as the fuel element is utilized. During the life of the fuel element, the decrease in excess reactivity is offset by the withdrawal of the neutron absorbing control rods, which permits the reactor to operate at its constant rated power levels.
The point in time denominated as the end of the fuel cycle is normally defined as that time when, with the control rods fully withdrawn and the overall system at full thermal load, the excess reactivity of the reactor core is zero. When these parameters are met, the end of the fuel cycle occurs, despite the fact that some residuum reactivity remains in the fuel elements. Of course, other parameters may be used to define at what point in time the end of the fuel cycle occurs.
When the reactor element operation nears the end of the fuel cycle, that is near the end of the useful life of the fuel element, even though there is still the residuum reactivity in the core, it has been observed that even total withdrawal of the control rods fails to sufficiently increase reactivity levels within the reactor to provide thermal output demanded by the electrical load. When reactivity decreases, the heat produced, or thermal output, within the core decreases. Decrease in thermal output of the reactor core due to decreased reactivity causes a concomitant lowering of water coolant temperatures. This in turn, decreases the amount of heat transfer in the steam generator. Thus, the thermal output of the steam generator is lowered, lowering the electrical output of the power plant as a whole.
It is well known to those skilled in the art that the density of the pressurized water coolant is dependent upon temperature of the coolant. As the temperature of the coolant decreases, the coolant becomes more dense. At the end of the fuel cycle, fissions of atoms within the fuel element still produce the residuum of reactivity noted above, but, the level of reactivity is not sufficient to maintain the rated thermal output of the reactor. The fissions produce both high energy and lower energy "thermal" neutrons. The thermal neutrons, although only a certain percentage of the neutrons released by the fission process, are important since they are required to enable the fission chain reaction to continue. Since the number of fissions, and therefore, the number of neutrons produced, is decreased at the end of the fuel cycle, and since only a predetermined percentage of neutrons produced are thermal neutrons, the reactivity, and hence the thermal output of the reactor, decreases.
However, since the pressurized water coolant becomes denser with a lower temperature, it is possible to increase the excess reactivity above the reference value by lowering coolant temperature. With coolant temperature lower, and the coolant more dense, more of the high energy neutrons produced by those fissions which provide the residuum reactivity will be slowed sufficiently to enable them to produce a fission. Thus, although the total number of neutrons released is not increased, the cooler, and therefore, denser, pressurized water coolant lowers the energy of enough high energy neutrons to provide enough thermal neutrons to sustain a fission chain reaction. In this way, the reactivity of the core is increased over the reference reactivity value, thus enabling the core to maintain its rated thermal output. The increase in excess reactivity thus offsets the loss of reactivity attendant upon the end of the fuel cycle.
However, decreases in coolant temperature have a deleterious effect upon pressure of the steam in the turbine side. In order to increase excess reactivity, without diminishing steam pressure, the prior art has to either insert a new fuel element or enrich the old fuel element. However, a further alternative which can maintain the reactivity at an excess level above the reference and the thermal output at its rated level, and yet still maintain the electrical output of the entire system within a predetermined close range of values to the rated electrical output is needed.